In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible ...with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.
A present topic of high interest in magnetic fusion is the “gap” between near-term and long-term concepts for high heat flux components (HHFC), and in particular for divertors. This paper focuses on ...this issue with the aim of characterizing the international status of current HHFC design concepts for ITER and describing the different technologies needed in the designs being developed for fusion power plants. Critical material and physics aspects are highlighted while evaluating the current readiness level of long-term concepts, identifying the design and R&D gaps, and discussing ways to bridge them.
Several advanced He-cooled W-alloy divertor concepts have been considered recently for power plant applications. They range in scale from a plate configuration with characteristic dimension of the ...order of 1
m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10
cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5
cm. The trend in moving to smaller-scale units is aimed at minimizing the thermal stress under a given heat load; however, this is done at the expense of increasing the number of units, with a corresponding impact on the reliability of the system. The possibility of optimizing the design by combining different configurations in an integrated design, based on the anticipated divertor heat flux profile, also has been proposed. Several heat transfer enhancement schemes have been considered in these designs, including slot jet, multi-hole jet, porous media and pin arrays. This paper summarizes recent US efforts in this area, including optimization and assessment of the different concepts under power plant conditions. Analytical and experimental studies of the concepts and cooling schemes are presented. Key issues are identified and discussed to help guide future R&D, including fabrication, joining, material behavior under the fusion environment and impact of design choice on reliability.
The dominating fraction of the power generated by fusion in the reactor is captured by neutron moderation in the blanket surrounding the plasma. From this, the efficiency of the fusion plant is ...predominated by the technologies applied to make electricity or hydrogen from the neutrons. The main blanket concepts addressed in this paper are advanced ceramic breeder concepts, dual coolant blankets as well as self-cooled liquid metal and Flibe blankets. Two important questions that are addressed are: (i) Can we draw a bottom line conclusion on the most promising concept(s)? (ii) What are the common issues to be resolved independently from individual design and layout proposals to define a feasible route towards advanced fusion reactors? For ceramic breeder concepts, a key issue in the long term could be the limitation of beryllium as the considered multiplier in terms of world sources and achievable temperature levels. For liquid metal blankets, attractive long-term visions have been developed but major technological challenges also exist for the in-vessel blanket technology and the corresponding sub-systems. The paper proposes a strategic conclusion derived from the review of blanket designs for advanced fusion reactors.
The pathway to the successful development of an attractive power plant implies a successful strategy in developing a number of key fusion systems forming part of the power plant. Here, an example ...strategy and pathway in developing a blanket system based on the lead–lithium alloy breeder is described as an illustration. A historical perspective of the development of the various lead–lithium based blankets is summarized and a thorough description of the key issues driving the blanket concept assessment is provided. A comprehensive assessment of the helium-cooled lead–lithium (HCLL) and dual-coolant lead–lithium (DCLL) blanket concepts is presented in the context of this development pathway and in helping guide the choice of the initial concept to be tested in ITER as part of an overall progressive development strategy.
The dual-coolant lead-lithium, or DCLL, blanket concept is of strong interest in the US fusion technology program. In the DCLL blanket, the flow channel insert (FCI) is a critical component. FCIs ...must have low electrical and thermal conductivity and be compatible with lead–lithium eutectic alloy (Pb–17Li) at elevated temperatures. FCIs must retain structural integrity and desirable properties even under irradiation and large temperature gradients during operation. FCIs must not fail in such a way that Pb–17Li enters the FCI and changes its electrical or thermal conductivity significantly. Another important issue for the DCLL is the development of a suitable tritium extraction from the Pb–17Li to achieve low tritium partial pressure, thus facilitating decisive tritium control. In this paper, the state of DCLL development in the US is presented including recent design modifications and results from recent R&D efforts. Such R&D includes the progress on development and property quantification of SiC/SiC composites and SiC foams as candidate FCI materials; Pb–17Li material capability and infiltration studies; simulations of MHD Pb–17Li flow characteristics and of resultant temperature distributions; and the analysis of FCI stress states based on these thermal loads. In addition, tritium extraction from Pb–17Li based on a vacuum permeator concept is shown to have the potential to achieve the desired tritium control. A discussion of DCLL optimization and unresolved DCLL issues and future R&D needs is also presented.
A number of different He-cooled divertor configurations have been proposed for magnetic fusion energy (MFE) power plant application. They range in scale from a plate configuration with characteristic ...dimension of the order of 1
m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10
cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5
cm. All these designs utilize tungsten or tungsten alloy as structural material. This paper considers the characteristics of the different divertor configurations and proposes the possibility of optimizing the design by combining different configurations in an integrated design based on the anticipated divertor heat flux profile.
ITER plasma-facing components Merola, Mario; Loesser, D.; Martin, A. ...
Fusion engineering and design,
12/2010, Letnik:
85, Številka:
10
Journal Article, Conference Proceeding
Recenzirano
The ITER plasma-facing components directly face the thermonuclear plasma and include the divertor, the blanket and the test blanket modules with their corresponding frames.
The divertor is located at ...the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimising the helium and impurity content in the plasma.
The blanket system provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel and external machine components. It consists of modular shielding elements known as blanket modules which are attached to the vacuum vessel. Each blanket module consists of two major components: a plasma-facing first wall panel and a shield block.
The test blanket modules are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to test blanket modules, each of them providing for the allocation of two breeding modules inserted in a steel frame and in front of a shield block.
An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas. ...The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D.
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